Text

Amends 39 & 10 to Licenses NPF-14 & NPF-22,respectively, Changing Tech Specs to Add Addl Rod Block Monitor Setpoint of 108%,change Labels on Listed Figures,Add Figure 3.2.3-1b & Update Figure 3.2.3-1
	ML17156A174
Person / Time
Site:	Susquehanna
Issue date:	04/23/1985
From:	Schwencer A; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML17156A175	List: ML17156A174; ML17156A176;
References
	NUDOCS 8504300060
	Download: ML17156A174 (31)
	v • d • e

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UNITEDSTATES NUCLEAR REGULATORY COMMtSSION WASHINGTON, D. C. 20555 PENNSYLVANIA'POWER 8I LIGHT COMPANY A L L

D E

N UUU IINUNNUNI~ITIN,UNIT I NDMNT F

C P

ING C

N Amendment No.. 39 License No. NPF-14 1.

The Nuclear Regulatory Commission (the Commission or the NRC) having found that:

A.

The application for the amendment filed by the Pennsylvania Power 8 Light Company, dated September 6,

1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C.

There is reasonable assurance:

(i) that the activities authorized

'by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-14 is hereby amended to read as follows:

(2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.

39 and the Environmental Protection Plan contained in Appendix 8, are hereby incorporated in the license.

The licensee shall operate the facilities in accordance with the Technical Specifications and the Environmental Protection Plan.

8504300060 850423 t

PDR ADOCK 05000387 P

PDRJ

3.

This amendment is.effective as of thirty (30) days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Enclosure:

Changes to the Technical Specifications Date of Issuance:

APR g g ~

A. Schwencer, Chief Licensing Branch No.

2 Division of Licensing

ENCLOSURE TO LICENSE AMENDMENT NO.

39 F

C I

N LIC N

N PF-14 D

K N

Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.

The revised pages are identified by Amendment number and contain a vertical line indicating the area of change.

REMOVE INSERT "3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-8 3/4 3-53 3/4 3-54 B 3/4 2-3 B 3/4 2-4 B 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-8 3/4 2-8.a 3/4 3-53 3/4 3-54 B 3/4 2-3 B 3/4 2-4 B 3/4 2-5

0

POMER DISTRIBUTION LIMITS 3/4. 2. 2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased simulated thermal power-upscale scram trip setpoint (S) and flow biased neutron flux-upscale control rod block trip setpoint (SRB) shall be established according to the following relationships:

~Ti S

Allowable Value S

< (G.SSW + 59K)T S <,(0.58M + 62K)T SRB < (0.58M + 50K)T SRB < (0.58M + 53K)T where:

S and S

B are in percent of RATED THERMAL PQMER, W = Lou/recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million lbs/hr, T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the MAXIMUM FRACTION OF LIMITING POMER DENSITY.

T is always less than or equal to 1.0.

P APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25K of RATED THERMAL POWER.

ACTION:

With the APRM flow biased simulated thermal power-upscale scram trip setpoint and/or the flow biased neutron flux-upscale control rod block trip setpoint less conservative than the value shown in the Allowable.Value column for S or S B, as above determined, initiate corrective action within 15 minutes and adjust 3 and/or S

to be consistent with the Trip Setpoint value within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL PIER to less than 25K of'ATED THERMAL POMER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREHENTS 4.2.2 The FRTP and the MFLPD shall be determined, the value of T calculated, and the most recent actual APRM flow biased simulated thermal power-upscale scram and flow biased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POMER increase of at least 15K of RATED THERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with MFLPD greater than or equal to FRTP.

d.

The provisions of Specification 4.0.4 are not applicable.

With HFLPD greater than the FRTP during power ascension up to 90K of RATED THERMAL POWER, rather than adjusting the APRH setpoints, the APRM gain may be adjusted such that APRM readings are greater than or equal to 100X times

MFLPD, provided that the adjusted APRM reading does not exceed 100K of RATED THERMAL POMER, the required gain adjustment increment does not exceed 10K of RATED THERMAL POMER, and a notice of the adjustment is posted on the reactor control panel.

SUS(UEHANNA " UNIT 1 3/4 2-5 Amendment No. QP

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR} shall be equal to or greater than the MCPR limit determined from Figure 3.2.3-1a or Figure 3.2.3. 1b, as applicable, times the Kf shown in Figure 3.2.3-2, provided that the end-of-cycle recirculation pump trip (EOC-RPT) system is OPERABLE per Specification 3.3.4.2 and the turbine bypass system is OPERABLE per Specification 3.7.8, with:

ave -

B

'A

'B xA = 0.86 seconds, control rod average scram insertion time limit to notch 39 per Specification 3.1.3.3, N~

tB = 0.688 + 1.65 [~

j (0.052),

1 1=1 n

Zi=l ave n

i=1 N-where:

n = number of surveillance tests performed to date in cycle, N. = number of active control rods measured in the i surveillance tests,

t. = average scram time to notch 39 of all rods measured in the i surveillance test, and N> = total number of active rods measured in Specification 4. 1.3.2. a.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25K of RATED THERMAL POWER.

SUS(UEHANNA - UNIT 1 3/4 2-6 Amendment No. 39

POWER DISTRIBUTION LIMITS I

LIMITING CONDITION FOR OPERATION Continued ACTION:

a 4 With the end-of-cycle recirculation pump trip system inoperable per Specification

3. 3.4. 2, oper ation may continue and the provisions of Specification 3.0.4 are not applicable provided that, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , MCPR is determined. to be greater than or equal to the MCPR limit as a

'unction of average scram time as shown in Figure 3.2.3-la or Figure 3.2.3. lb, as applicable, EOC-RPT inoperable curve, times the Kf shown in Figure 3.2.3-2.

b.

With the turbine bypass system inoperable per Specification 3.7.8, opera-tion may continue and the provisions of Specification 3.0.4 are not appli-cable provided that within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , MCPR is determined to be greater than or equal to the MCPR limit as a function of average scram time as shown in Figure 3.2.3-la or Figure 3.2.3. 1b, as applicable, turbine bypass inoper-able curve.

times the Kf shown ln Figure 3 2.3-2 c.

With MCPR less than the applicable MCPR limit determined from Figure 3.2.3-1a

- or Figure 3.2.3. lb, as applicable, and Figure 3.2.3-2, initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURYEILLA'NCE RE UIREMENTS 4.2.3 MCPR, with a.

x = 1.0 prior to performance of the initial scram time measurements for the cycle in accordance with Specification 4. 1.3. 2, or b.

x as defined in Specification 3.2.3 used to determine the limit within.72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram time surveillance test required by Specification 4.1.3.2,

'hall be determined to be equal to or greater than the applicable MCPR limit determined from Figure 3. 2.3-1a or Figure 3. 2. 3. lb, as applicable, and Figure 3.2.3-2:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion *of a THERMAL POWER increase of at least 15K of RATED THERMAL POWER, and c.

Initially and at -least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL ROD PATTERN FOR MCPR.

d.

The provisions of Specification 4.0.4 are not applicable.

SUSQUEHANNA " UNIT 1 3/4 2-7 Amendment No. 39

MINIMUMCRITICAL POWER RRIO (MCPR)

YERSUS 7 AT RATED FLOW*

1AO

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1.24 1.22

'1,20 Curve A:

Curve B:

Curve C:

J Main Turbine Bypass Inoperable; EOC RPT Operable EOC RPT Inoperable; Main Turbine Bypass Operable EOC-RPT and Qafn Turbine Bypass'Operable RBM AT 1067o PER TABLE 3.3.6-2 RQURE $.2.3-1a SUSQUEHANNA - UNIT 1 3/4 2-8 Amendment No. 39.

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MINIMUMCRITICAL POWER RATIO (MCPR)

VERSUS v AT. RATED<.FLOVl*

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1.42 1.40 1.38 1.36

~ 1034 1.32 1.30 1.28 1.26 1.24 1.22 1.20

.Curve A:

Curve B:

Curve C:

Main Turbine Bypass Inoperable; EOC RPT Operable EQC RPT Inoperob4r, I4ain Turbine Bypass Operable EOC RPT and lAatn Turbine Bypass Operable

~ RBM AT 1087o PER TABLE 3.3.6-2 SUSQUEHANNA - UNIT 1

\\ ~

RQURE.3.2.3-1b

'/4 2-Ba Amendment No.'39

.i)'

TABLE 3.3.6-1 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION ACTION ACTION 60 Declare the RBM inoperable and take the ACTION required by Specification 3.1.4.3.

ACTION 61 With'-the'umber of OPERABLE Channels:

a.

One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 7 days or place the inoperable channel in the tripped condition within the next hour.

b.

Two or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 62 With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

NOTES With THERMAL POWER > 30K of RATED THERMAL POWER.

With more than one control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.

1 or 3.9.10.2.

a.

The RBM shall be automatically bypassed when a.peripheral control rod is selected or the reference APRM channel indicates less than 30K of RATED THERMAL POWER.

b.

This function shall be automatically bypassed if detector count rate is

> 100 cps or the IRM channels are on range 3 or higher.

c.

This function is automatically bypassed when the associated IRM channels are on range 8 or higher.

d.

This function is automatically bypassed when the IRM channels are on range 3 or higher.

e.

This function is automatically bypassed when the IRM channels are on range l.

SUS(UEHANNA - UNIT 1 3/4 3-53 Amendment No.49

TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS

~

CA C/1 AD m

I a.

b.

C.

Upscale 1) 106X 2) 108X¹ Inoperative Downscale TRIP FUNCTION 1.

ROD BLOCK MONITOR TRIP SETPOINT

< 0.66 W + 40X

< 0.66 W+ 42X, NA.

> 5/125 divisions of full scale ALLOWABLE VALUE

< 0.66 W+ 43X

< 0.66 W+ 45X NA

> 3/125 of divisions full scale 2.

3.

APRH a.

Flow Biased Neutron Flux - Upscale b.

Inoperative c.

Downscale d.

Neutron Flux - Upscale Startup SOURCE RANGE MONITORS a.

Detector not full in b.

Upscale c.

Inoperative d.

Downscale

<.0.58 W + 50X" HA

> 5X of RATED. THERMAL POWER

< 12X of RATED THERMAL POWER NA

< 2x10 cps HA 0 7 cps Itchy'

< 0.58 W + 53X" t

NA

> 3X of RATED THERMAL POWER

< 14X of RATED THERMAL POWER NA<4x10 cps NA

> 0.5 cps"",

4.

INTERMEDIATE RANGE MONITORS a.

Detector not full in b.

Upscale Inoperative d.

Downscale 5.

SCRAM DISCHARGE VOLUME NA

< 108/125 divisions of full scale NA

> 5/125 divisions of full scale NA

< 110/125 divisions of full scale NA

> 3/125 divisions of full scale a.

Water Level - High

< 44 gallons 6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.

Upscale

< 108/125 divisions of full scale b.

Inoperative HA c.

Comparator

< 10X flow deviation

< 44.gallons

< ill/125 divisions of full scale HA

< 11X flow deviation O

e Average Power Range Monitor rod block function is varied as a function of recircul'ation loop flow (W).

The trip setting of this function must be maintained in accordance with Specification 3.2.2.

"*Provided signal-to-noise ratio is >2.

Otherwise, 3cps as trip setpoint and 2.8cps for allowable value.

¹May be used when the associated HCPR requirements in Specification 3.2.3 are satisfied.

Bases Table B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF" COOLANT ACCIDENT ANALYSIS Plant Parameters; Core THERMAL POWER.....................

3439 Mwt" which corresponds to 105K of rated 'steam flow Vessel Steam Dome Pressure..............

1055 psia Design Basis Recirculation Line Break Area for:

a.

Large Breaks 4.153 ft b.

Small Breaks 1.0 ft to 0.02 ft Fuel Parameters:

Vessel Steam Output....................

14. 15 x 10 ibm/hr which cor-6 responds to 105K of rated steam flow FUEL TYPE Initial Core

'FUEL BUNDLE GEOMETRY 8 x 8

13. 4 1.4

l. 18 PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT

~ AXIAL CRITICAL GENERATION RATE PEAKING POWER (kw/ft)

FACTOR RATIO A more detailed listing of input of. each model and its source is presented in Section II of Reference 1 and Section 6.3 of the FSAR.

"This power level meets the Appendix requirement of 102K.

The core heatup calculation assumes a bundle, power consistent with operation of the highest powered rod at 102K of its Technical Specification LINEAR HEAT GENERATION RATE limit.

SUSQUEHANNA - UNIT 1 B 3/4 2-3

0 POWER DISTRIBUTION LIMITS=

BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit HCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.06, and an analysis of abnormal operational transients.

For any abnormal operating transient.analysis evalua-tion with the initial condition of the reactor being at the steady state...'.

operating limit, it-4s rejuired ghat the resulting MCPR does not decrease below the Safety Limit HCPR ai any time during the transient assuming instrument trip setting given in Specification 2.2.

'o assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POMER RATIO (CPR).

The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant.

temperature decrease.

The limiting transient yields the largest delta MCPR.

Mhen added to the Safety Limit HCPR of 1.06, the required minimum operating limit MCPR of Specification 3.2.3 is obtained and presented in Figure 3. 2.3-1a and 3.2.3-1b.

Mhen the less operationally limiting Rod Block Monitoring trip setpoint

(.66M+ 42K from Table 3.3.6"2) is used, the more limiting MCPR curve Figure 3.2.3-lb is applicable due to a larger delta MCPR from the limiting Rod Withdrawal Error (RQE) transient.. Figure 3.2.3-1a is applicable when the Rod Block Honitor trip setpoini (.66M + 40K f'rom TAble 3.3.6-2) is used.

The evaluation of a given transient begins with the system initial parameters shown in FSAR Table 15.0-2 that are input to a GE-core dynamic behavior transient computer program.

The code used to evaluate pressurization events is described in NEDO-24154 and the program used in nonpressurization events (2) is described in NEDO-10802 The outputs of this program along with the initial HCPR form the input foi further analyses of the thermally limiting bundle with the single channel transient thermal hydraulic TASC code described in NEDE-25149 The principal result of this evaluation is the reduction in HCPR caused by the transient.

The purpose of the Kf factor of Figure 3.2.3-2 is to define operating limits at other than rated core flow conditions.

At less than 100K of rated flow the required MCPR is the product of the MCPR and ihe Kf factor.

The Kf factors assure that the Safety Limit MCPR will not be violated during a flow increase transient resulting from a motor-generator speed control failure.

The Kf factors may be applied to both manual and automatic flow control modes.

The Kf factor values shown in Figure 3.2.3-2 were developed generically and are applicable io all BWR/2, BWR/3 and BMR/4 reactors.

The Kf factors were derived using the flow control line corresponding to RATED THERMAL POWER at rated core flow.

For the manual flow control mode, the Kf factors were calculated such thai for the maximum flow rate, as limited by the pump scoop tube set point and the corresponding THERMAL POMER along the rated flow control line, the limiting SUStlUEHANNA - UNIT 1 B 3/4 2"4 Amendment No..

39

POMER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER-RATIO (Continued) bundle's relative power was adjusted until the MCPR changes with different core flows.'he ratio of the HCPR calculated at a given point of core flow, divided by the operating limit MCPR, determines the Kf.

For operation in the automatic flow control mode, the same procedure was.

employed except the initial power distribution was established such that the HCPR was equal to the oper ating limit MCPR at RATED THERMAL POWER and rated flow.

The Kf factors shown in Figure 3.2.3-2 are conservative for the General Electric Plant operation because the operating limit MCPRs of Specifica-tion 3.2.3 ar'e greater than the original 1.20 operating limit MCPR used for the generic derivation of Kf.

At THERMAL POWER levels less than or equal to 25K of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small.

For all designated control rod patterns which may be employed at this point, operating plant experience indi-cates that the resulting MCPR value is in excess of requirements by a considerable margin.

During initial start-up testing of the plant, a

MCPR evaluation will be made at 25K of RATED THERHAL POWER. level with minimum r'ecirculation pump speed.

The HCPR margin will thus be demonstrated such that future HCPR evaluation below this power level will be shown to be'unnecessary.

The daily requirement for calculating HCPR when 'THERMAL 'POWER is greater than or equal to 25K of RATED THERMAL POMER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.

The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.

3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate

{LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.

References:

2.

3.

General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566, November 1975; R.

B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, February 1973 {NED0-10802).

gualification of the One Dimensional Core Transient Model For Boiling Mater Reactor, NED0-24154, October 1978.

TASC-Ol-A Computer Program For the Transient Analysis of a Single

-Channel, Technical Description, NEDE-25149, January 1980.

SUSQUEHANNA " UNIT 1 B 3/4 2-5 Amendment No.

39

>1 I'3 I

P

~S Rfgy Cy

~i nO IlA e

0' V/+~

gO

++*++

. ~

UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASMINGTON,D. C. 20555 2ENNSYLVANIA POWER 5 LIGHT COMPANY D CKET N

. 50-3 8 EEEQIIEEAIINAN~T QN. IINIT I D

N FCILITY PR NGLCNS Amendment No.

10 License No. NPF-22 1.

The Nuclear Regulatory Commission (the Commission or the NRC) having found that:

A.

The application for the amendment filed by the Pennsylvania Power and Light Company, dated September 6, 1984, complies with the standards and requirements of the Atomic Energv Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.

C.

D.

The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; There is reasonable'ssurance:

(i) that the activities authorized by this amendment'can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; The<issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-22 hereby amended to read as follows:

(2)

Technical.

S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.

10, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

The licensee shall operate the facility in accordan'ce with the Technical Specifications and the Environmental Protection Plan.

3.

This amendment is.effective as of thirty (30) days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Enclosure:

Changes to the Technical Specifications Date of Issuance:

APE 8 3 $85 A. Schwencer, Chief Licensing Branch No.

2 Division of Licensing

ENCLOSURE TO LICENSE AMENDMENT NO.

10 K

N.5-3 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.

The revised pages are identified by Amendment number and contain a vertical line.indicating the area'of change.

REMOVE 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-8 3/4 3-53 3/4 3-54 B 3/4 2-3 B 3/4 2-4 B 3/4 2-5 INSERT 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-8 3/4 2-8.a 3/4 3-53 3/4 3-54 B 3/4 2-3 B 3/4 2-4 B 3/4 2-5

, POWER DISTRIBUTION LIMITS 3/4. 2. 2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased simulated thermal power-upscale scram trip setpoint (S) and flow biased neutron flux-upscale control rod block trip setpoint (SRB) shall be established according to the following relationships:

I Tri Set oint Allowable Value S <

0.58W + 59K)T S

< (0.58W + 62K)T SRB

< (0.58W + 50m)T SRB (0.58W + 53K)T where:

S and S

B are in percent of RATED THERMAL POWER, H = LooPrecirculation

.flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million lbs/hr, T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY.

T is

'lways less than or equal to 1.0.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or

~q1 TER TIIERRAE 1 ER.

ACTION:

With the APRM flow biased simulated thermal power-upscale scram trip setpoint and/or the flow biased neutron flux-upscale control rod bl'ock trip setpoint less conservative than the value shown in the Allowable Value column for S or 5

, as above determined, initiate corrective action within 15 minutes and a) ust S and/or S a to be consistent with the Trip Setpoint value* within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POHER to less than 25A'f RATEO THERMAL POWER within RB the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 4.2.2 The FRTP and the MFLPD shall be determined, the value of T calculated, and the most recent actual APRM tlow biased simulated thermal power-upscale scram and flow biased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required:

r a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15K of RATED THERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with MFLPD greater than or equal to FRTP.

d.

The provisions of Specification

4. 0.4 are not applicable.

"With MFLPD greater than the FRTP during power ascension up to 90K of RATED THERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that APRM r'eadings are greater than or equal to 100K times MFLPD, provided that the adjusted APRM reading does not exceed 100K of RATED THERMAL POWER, the required gain adjustment increment does not exceed lOX of RATED THERMAL POWER, and a notice of the adjustment is posted on the reactor control panel.

SUSQUEHANNA " UNIT 2 3/4 2-5

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 0

3.2.3 The MINIMUM CRITICAL POWER RATIO (HCPR) shall be equal to or greater than the MCPR limit determined from Figure 3.2.3-1a or Figure 3.2.3-lb, as applicable, times the Kf shown in Figure 3.2.3-2, provided that the end-of-cycle recirculation pump trip (EOC-RPT) system is OPERABLE per Specification 3.3.4.2 and the turbine bypass system is OPERABLE per Specification 3.7.8,.with:

ave -

B A

B xA = 0.86 seconds, control rod average scram.insertion time limit to notch 39 per Specification 3. 1.3.3, Nl zB = 0.688 + 1.65 [

l (0.052)

Z l=1 n

Z 1=1 ave n

1=1 N.

where:

n = number of surveillance tests performed to date in cycle, N. = number of active control rods measured in the i surveillance tests,

.x. = average scram time to notch 39 of all rods measured in the i surveillance test, and N> = total number of active rods measured in Specification 4. $. 3.2. a.

APPLICABILITY'PERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25K of RATED THERMAL POWER.

SUS(UEHANNA " UNIT 2 3/4 2-6 Amendment No. l0

0 POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION Continued)

ACTION:

b.

C, With the end-of-cycle recirculation pump trip system inoperable per Specification 3.3.4.2, operation may continue and the provisions of Specification 3.0.4 are not applicable provided that, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , MCPR is determined to be greater than or equal to the MCPR limit as a

'unction of average scram time as shown in Figure 3.2.3-la or Figure 3.'2.3-1b, as applicable, EOC-RPT inoperable curve, times the Kf shown in Figure 3.2.3-2.

With the turbine bypass system inoperable per, Specification 3.7.8, operation may continue and the provisions of Specification 3.0.4 are not applicable provided that within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , MCPR is determined to be greater than or equal to the MCPR limit as a function of average scram time as shown in Figure 3.2.3-la or Figure 3.2.3-1b, as applicable,

, turbine bypass inoperable curve, times the Kf shown in Figure 3.2.3-2.

With MCPR less than the applicable MCPR limit determined from Figure 3.2.3-la or Figure 3.2.3-lb, as applicable, and Figure 3.2.3-2, initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .or reduce THERMAL POWER to less than 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 4.2.3 MCPR, with ao b.

x = 1.0 prior to performance of the initial scram time measurements for the cycle in accordance with Speci ficati on 4. 1.3. 2, or t as defined in Specification 3.2.3 used to determine the limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram time surveillance test required by Specification 4. 1.3.2, c.

The provisions of Specification 4.0.4 are not applicable.

shall be determined to be equal to or greater than the applicable MCPR determined from Figure 3.2. 3-la or Figure 3. 2. 3-lb, as applicable, and 3.2.3-2:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , limit Figure b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15K of RATED THERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL ROD PATTERN FOR MCPR.

d.

The pr'ovisions of Specification 4.0.4 are not applicable.

SUSQUEHANNA " UNIT 2 3/4 2-7 Amendment No. l0

I

. PIMUM CRITICAL POWER ETIO (MCPR)

VERSUS v AT RATED FLOV/*

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Curve C:

Main Turbine Bypass Inoperable; FOG RPT Operable ROC RPT Inoperable; Main Turbine Bypass Operable COC RPT And Main Tvrbine Bypass Operable

+ RBM AT 1087o PER TABLE 3.3.6-2 RGURE 3.2.3-1b SUSQUEHANNA - UNIT 2 3/4 2"Sa Amendment No 1 0

Ã

TABLE 3. 3. 6-1 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION ACTION 60 ACTION 61 ACTION 62 ACTION Declare the RBM inoperable and take the ACTION required by Specification 3.1. 4.3.

With the number of OPERABLE Channels:

a.

One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable

., channel to OPERABLE status within 7 days or place the inoperable channel in the tripped condition within the next hour.

b.

Two or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

NOTES With THERMAL POWER

> 30K of RATED THERMAL POWER.

With more 'than one control rod withdrawn.

Not applicable to control, rods removed per Specification 3.9.10.1 or 3.9.10.2.

(a)

The RBM shall be automatically bypassed when a peripheral control rod is selected '-or the reference APRM channel indicates less than 30K of RATED THERMAL, POWER.

(b)

This function shall be automatically bypassed if detector count rate is

> 100 cps or the IRM channels are on range 3 or higher.

(c)

This function is automatically bypassed when the associated IRM channels are on range 8 or higher.

(d)

This function is automatically bypassed when the IRM channels are on range 3 or higher.

(e)

This function is automatically bypassed when the IRM channels are on range 1.

SUSQUEHANNA " UNIT 2 3/4 3"53

TABLE 3;3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION 1.

ROD BLOCK MONITOR TRIP SETPOINT ALLOWABLE VALUE a 0 b.

C.

Upscale 1) 106X 2) 108'noperative Downscale 0.66 W + 40X 0.66 W + 42X

< 0.66 W + 43X

< 0.66 W + 45X.

NA.

NA

> 5/125 divisions of full scale

> 3/125 of divisions full scale 2.

3.

APRH a.

Flow Biased Neutron Flux - Upscale b.

Inoperative c.

Downscale d.

Neutron Flux - Upscale Startup SOURCE RANGE MONITORS a.

Detector not full in b.

Upscale c.

Inoperative d.

Downscale

< 0.58 W + 50X" NA

> 5X of RATED THERMAL POWER

< 12X of RATED THERMAL POWER NA

< 2 x 10 cps NA 0 7 cpsA I

< 0.58 W + '53X" NA

> 3X of RATED THERMAL POWER '

14X of RATED THERMAL POWER NA

< 4 x 10 cps NA

> 0.5 cps""

4.

INTERMEDIATE RANGE MONITORS

'a.

b.

C.

d.

Detector not full in Upscal e Inoperative Downscale NA

< 108/125 divisions of full scale NA

> 5/125 divisions of full scale NA

< 110/125 divisions of full scale NA

> 3/125 divisions of full scale 5.

SCRAM DISCHARGE VOLUME a.

Water Level - High

< 44 gallons 6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW

< 44 gallons

a. 'pscale b.

Inoperative'.

Comparator

< 108/125 divisions of full scale

< ill/125 divisions of full scale NA NA

< 10X flow deviation

< 11X flow deviation The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow (W).

The trip setting of this function must be maintained in accordance with Specificgtion 3.2.2.

"*Provided signal-to-noise ratio is > 2.

Otherwise, 3 cps as trip setpoint and 2.8 cps for allowable value:

O'Hay be used when the associated HCPR requirements in Specification 3.2.3 are satisfied.

Bases Table B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS Plant Parameters';

Core THERMAL POQER.....................

3439 Mwt" which corresponds to 105K of rated steam flow Vessel Steam Output...........

14.15 x 10 ibm/hr which cor-6 responds to 105K of rated steam flow Vessel Steam Dome Pressure..............

1055 psia A

Design Basis Recirculation Line Break Area for:

a.

Large Breaks 4.153 ft b.

Small Breaks, 1.0 ft to 0.02 ft Fuel Parameters:

FUEL TYPE Initial Core FUEL BUNDLE GEOMETRY PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL GENERATION RATE PEAKING POMER (kw/ft)

. FACTOR RATIO 8 x 8

13. 4 1.4 1.18 A more detailed listing of input of'ach model and its source is presented in Section II of Reference 1 and Section 6.3 of the FSAR.

"This power level meets the Appendix requirement of 102K.

The core heatup calculation assumes a bundle power consi.stent with operation of the highest powered rod at 102K of its Technical Specification LINEAR HEAT GENERATION RATE limit.

SUS(UEHANNA - UNIT 2 B 3/4 2-3

0 POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3. 2.3 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.06, and an analysis of abnormal operational transients.

For any abnormal operating transient analysis evalua-tion with the initial condition of the reactor being at the steady state operating limit, it-is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR).

The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The limiting transient yields the largest delta MCPR.

When added to the Safety Limit MCPR of 1.06, the required minimum operating limit MCPR of Specification 3.2.3 is obtained and presented in Figure 3.2.3-la and 3.2.3-1b.

When the less operationally limiting Rod Block Monitoring trip setpoint

(.66W + 42K from Table 3.3.6-2) is used, the more limiting MCPR curve Figure

3. 2. 3-lb is applicable due to a larger delta MCPR from the limiting Rod With-,

drawal Error (RWE) transient.'igure

3. 2.3-la is applicable when the Rod Block Monitor trip setpoint

(.66W + 40K from Table 3.3.6-2) is. used.

The evaluation of a given transient begins with the system initial parameters shown in FSAR Table 15.0-2 that are input to a GE-core dynamic behavior transient computer program.

The code used to evaluate pressurization events is described in NEDO-24154 and the program used in nonpressurization events is described in NEDO-10802 The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle with the single channel transient thermal hydraulic TASC code described in NEDE.-25149 The principal result of this evaluation is the reduction in MCPR caused by the transient.

The purpose of the Kf factor of Figure 3.2.3-2 is to define operating limits

'at other than rated core flow conditions.'t less than 100K of rated flow the required MCPR is the product of the MCPR'nd the Kf factor.

The K

factors assure that the Safety Limit MCPR will not be violated during a flow increase transient resulting from a motor-generator speed control failure.

The Kf factors may be applied to both manual and automatic flow control modes.

The K

factor values shown in Figure 3.2.3-2 were developed generically and are apP feeble to all BWR/2, BWR/3 and BWR/4 reactors.

The K

factors.were derived using the flow control line corresponding to RATED THERMA[ POWER at rated core flow.

SUS(UEHANNA - UNIT 2 B 3/4 2-4 Amendment No.10

, ~

POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)

For the manual flow control mode, the Kf factors were calculated such that for the maximum flow rate, as limited by the pump scoop tube set point and the corresponding THERMAL POWER along the rated flow control line, the limiting bundle's relative power was adjusted until the MCPR changes with different core flows.

The ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR, determines the Kf.

For operation in the automatic flow control mode, the same procedure was employed except the initial power distribution was established such that the MCPR was equal to the operating limit MCPR at RATED THERMAL POWER and rated flow.

The Kf factors shown in Figure 3.2.3-2 are conservative for the General Electric Plant operation because the operating limit MCPRs of Specifica-tion 3.2.3 are greater than the original 1.20 operating limit MCPR used for the generic derivation of Kf.

At THERMAL POWER levels less than or equal to 25'f RATED THERMAL POWER, the'reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small.

For all designated control rod patterns which may be employed at this point, operating plant experience indi-cates that the resulting MCPR value is in excess of requirements by a

'- considerable margin.

During initial start-up testing of the plant, a

MCPR evaluation will be made at 25K of RATED THERMAL POWER level with minimum recirculation pump speed.

The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary.

The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25K of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.

The~requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.

3/4. 2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.

References:

2.

3.

4, General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566, November 1975.

R.

B. Linford, Analytical Methods of Plant Transient Evaluations, for the GE BWR, February 1973 (NED0-10802).

gualification of the One Dimensional Core Transient Model For Boiling Water.Reactor, NED0-24154, October 1978.

TASC 01-A Computer Program For the Transient Analysis of a Single"

Channel, Technical Description, NEDE-25149, January 1980.

SUSQUEHANNA - UNIT 2 B 3/4 2"5 Amendment No. 10

I

ML17156A174

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